ML20247C558
| ML20247C558 | |
| Person / Time | |
|---|---|
| Issue date: | 09/06/1989 |
| From: | Grimsley D NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Johnson G SWIDLER & BERLIN |
| References | |
| FOIA-89-244 NUDOCS 8909140041 | |
| Download: ML20247C558 (3) | |
Text
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ASPONSE TYPE RESPON E-O REEDO OF
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I le^"I^t INFORMATION ACT (FOIA) REQUEST SEP 6 1989 DOCKET NUMBER ($1(# apphcebier Rt QUt Stir dl D b* O l*
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PART t.- AGENCY RECORDS hELEASED Ol[NOT LOCATED (See checked t>oacs)
No agency records sub;est to the request have been located.
l' No additional agency records subject to the request have been located.
Requested records are available through another public distribution program, See Comments Section.
l Agency records subsect to the request that are identified on Appendiales) are already eveilable for pubhc mspection and copymg in the NRC Public Document Room 2120 L Street. N.W., Washington. DC 20555.
Agency records subsect to the request that are identified on Appendiales) me being made available for public mapection and copying in the NRC Publec Document Room, 2120 L Street, N.W., Washington, DC in a folder under this FOIA number and requester nome.
The nonproprietary version of the proposet(s) that you agreed to accept m a telephone conversation with a member of my staff is now being made evallable for public in:pection and copying at the NRC Pubhc Document Room 2120 L Street. N.W., Washington. OC, m o folder undes ther QiA number and sequester n* me.
Agency records subject to the request that are identsfiedon Appendiales) may be inspected and copied at the NRC Local Pubhc Document Room identified h the Comments Section.
Enctosed is information or how you may obtain eccess to end the charges for copymg records placed in the NRC Pubhc Document Room. 2120 L Street, N W.,
Wrshington. DC.
Agency rs, cords subject to the request are enclosed. M Records subiect to the request have teen referred to another Federal agencyhes) for review and detect response to you.
NRC St fees totahnD 8 !
In view of NRC's response to this request, no further action is berng taken on appeal letter dated No.
PART 11. A-INFORMATION WITHHELO FROM PUBLIC DISCLOSURE Certam information in the requested reco*ds is being withheld from pubhc disclosure pursuant to the esemptions desenbed m and for the reasons stated m Port II, sections B. C. and D. Any released portions of the documents for which only part of the record is being wethheld are bemp made availebte for pubhc mspection and i
copymg in the NRC Public Document Room. 2120 L Street, N W., Washington, DC.m a folder under this FOIA number and requester name.
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11.
Inspection Reports
' 50-293/86-14 (Addresses 5/16/86 meeting)
. 50-293/86-22 (Addresses 6/12/86 meeting) 50-293/86-26 (Addresses 7/30/86 meeting) 2.
5/16/86 W.'Kane' W. Harrington, BECo letter Sucject:
Reg'On I, NRC Boston Edison-Response to CAL 86-10 3.
6/10/86 H. (ister W. Harrington,- BECo Letter
Subject:
Reg'on I, NRC_
Management Meeting on CAL 8610 4,
7/18/86
' 05'.::, Region I Notice of Licensee Meeting No.186-084 5.
8/29/86 W. rarrington, BECo USNRC, Region I Licensee lette'r (PDR Accession No. 8609050027) 1 f
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REEORDS MAINTAINED IN THE POR UNDER THE ABOVE REQUEST NUMBER DATE ORIGINATOR RECIPIENT DESCRIPTION 1*
6/16/86 W. D. Harrington Dr. Murley, Region I Letter
Subject:
BECo NRC Second Response to NRC CAL 86-10, Regarding the Events which occurred on April 4, 11-12, 1986 at Pilgrim Nuclear Power Station (77 pages)
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June 16, 1986 BECo Ltr #86-079 j
Dr.' Thomas E. Murley Regional Administrator it 9. Nuclear Regulatory Commission 631. P
.-k Avenue - Region 1 King of Prussia, PA 19406
Subject:
Second Response to NRC Confirmatory Action Letter #86-10, Regarding the Events Which Occurred on April 4, 11-12, 1986 at Pilgrim Nuclear Power Station
References:
(A) NRC CAL 86-10 dated April 12, 1986 (B) BECo Response to CAL 86-10 dated May 15, 1986 (C) NRC Inspection Report 86-17 dated Ny 16, 1986, Documenting the NRC's Augmented Inspection Team's findings, Conclusions, and Recommendations (D) NRC " Request for Additional Information" letter dated my 16, 1986
Dear Dr.. Murley:
This letter provides the additional information requested in your. letter dated May 16, 1986' a,nd in a May 19th meeting held at the Region 1 office.
Answers to the specific questions of the letter are included as Attachment (1) to this letter, and the enclosed Table of Contents will provide a cross-reference for other information requested by members of your staff.
We trust that the contents of this submittal combined with information provided in our preliminary response dated May 15, 1986 will provide information adequate to address the requirements of Confirmatory Action Letter 86-10.
,Should you have further questions concerning these matters, please do not hesitate to contact me, Sincerely, i
W. D. Harringt JC/vep j
i Attachments Wew? r2A *h&
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TABLE Or' CONTENTS
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Attachments TITLE PAGE 1
BECo Response to NRC Questions as Al-1 thru A1-24 Presented in NRC Letter dated May 16, 1986 2
BfCo Responses to NRC Questions as A2-1 thru A2-8 Presented at May 19, 1986' Meeting at Region 1 Office 3
Revised PNPS Procedure 2.3.2.1 A3-1 thru A3-12 Aldrm Procedure Panel 903 Left 4'
Temporary PNPS Procedure TP 86-85 A4-1 thru A4-9 RHR Intersystem Leakoge Assessment and Controlled Leakoff 5
. Temporary PNPS Procedure TP 86-84 AS-1 thru A5-6 RHR Discharge Piping Venting Procedure 6.
Safety Evaluation SE 1959 A6-1 t,hru A6-6 Establishment of a Controlled Leak in the RHR System per Temporary Modification 86-22 7
Plan and Schedule Details Regarding A7-1 thru A7-3 Various Long Term Actions
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ATTACHMENT 1 s
BLCo Responses to NRC Questions As Presented In NRC Letter Dated May 26, 1986 1
RHR' What was the rationale for not. refurbishing NRC Question; a.
both the 28B and 298 RHR valves and how will you control the normally' closed injection valve in the "B" RHR loop in the future to
~
isolate the: RHR system?
BECo Response:
We decided to refurbish only the 288 valve for the following reasons:
1.
The "as found" leak rates were. low and well within Type C LLRT limits.
Extensive maintenance experience with the 29B
. valve indicated that.there was very little d'$
5 probability of improving upon its."as found" Y,() ih' "44 leak rate, A
g,'
1he 288 valve was the "normally closed" injection valve when the RHR pressurization problem began; therefore, a change in this valve's leak rate had occurred.
4 The 288 valve' is used to throttle flow and therefore, has the highest probability of experiencing seating surface wear.
The normally closed injection valve in the "B" RHR Loop will be controlled by system lineup procedures, It is expected that the MO 1001-28B valve will be the normally closed isolation valve, upon startup from the present outage.
NRC Question:
b.
What is the scope of the planned inspection for 288 valve?
BECo Response:
An inspection of the disassembled MO 1001-280 valve internals was performed by BECo Maintenance and Quality Control personnel. This consisted of a visual examination of the valve seat and disc and a bluing check of the disc seat; The NRC Resident inspector witnessed the valve disassembly. The valve was local leak rate tested with air to 10CFR50 Appendix J criteria and with water at 950 psig as part of postwork testing.
Inspection and testing results are presented in Attachment (2) to this letter.
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' Which recommendations in Artschment 4 to your.
NRC Quastion:
c.
letter cpply to both RHR loops? For exceple, will pressure and temperature instrumentation be installed'on both loops?
BECo Response:
The following recommendations from our prior response letter (Reference B) apply to both RHR loops:
B.1, C.1..b, C.1.c, C. 2. a, C. 3. a, C. 3 b, C. 4. a, C. 4. b, C. 5. a, and C. 6.
A table explaining these items is included later in this response (Page A1-24).
More specifically. pressure and
- temperature monitoring devices will be installed on both loops of the.RHR system.
NRC Question; d.
Are baseline torque measurements on the RHR motor operated injection valves to be taken?
BECo Response:
No.
However, closure seating current as well as motor running current' are recorded as part of post-maintenance testing of the RHR motor operated injection valves in accordance with procedure 3.M.4-10.
These current measurements a:-e indicative of the operational condition of this motor operator, What method will be used to control RHR NRC Question:
e.
system leakoff? How will the amount of leakage into the low pressure portion of the RHR system be measured? If a bypass valve is used, when will it be opened or closed? Will any administrative procedures be put in place to monitor and control letdown to the suppression pool and suppression pool level?
A safety evaluation should be submitted to the NRC which evaluates the leakoff method.
BECo Response:
RHR system leakoff will be controlled through throttling open a bypass valve around the RHR pump discharge check valve.
Prior to opening this valve, the RHR intersystem leakage will be quantified by maintaining a pressure in the RHR system above the keep fill pressure (>125 psig) and measuring the unit volume of fluid leakage per time to maintain the set r
pressure.
The RHR pump discharge check valve bypass valve will then be throttled open to maintain the same pressure as used in the measurement step. This valve will then be secured in that position with tie wraps.
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The measured leakagd will be controlled procedurally.
Continued station oepration will not be permitted above.1.0 gpm withouth an additional Nuclear Engineering Department evaluation. _Once the throttled valve has been opened it will remain open as a controlled leak path. Subsequent RHR high
. discharge pressure alarm while in normal power operation will indicate that RHR intersystem leakage has exceeded the procedural limit and that additional corrective actions are to be initiated. Such corrective actions will. include quantification of the leakage, temperature monitoring, and evaluation of acceptability of the leakage. A unit shutdown will take place if the 10 gpm flow rate is exceeded.
Since post-work testing of.the MO 1001-288 valve did not exhibit.an improvement in leak tightness the RHR leakoff will be used upon unit startup.
In the event that RHR high pressure alarms recur, the controlled leakoff will be implemented. The alarm procedure has been revised to key the described actions and a procedure has been prepared to implement the controlled leakoff. The procedures are included as Attachments (3) and (4) to this response.
Administrative procedures are already in place to monitor and control suppression pool level per technical specifications.
A ' safety evaluation has been prepared for the controlled leakoff and is included as Attachment (6) to this response.
Training of the operating staff on the RHR intersystem leakage issue will be conducted on shift concurrent with plant startup.
NRC Question:
f.
What type of test (water or air) will be used to check the pressure drop across the 68 check valves? What test pressure will be used? What acceptance criteria will be used? Will this test be conducted at refueling outages or af ter each time the check valves are cycled?
BECo Response:
Water testing of the injection check valve will be used as the preferred method for testing pressure drop capability of the 1001-6BA and 1001-68B check valves. The test pressure will be approximately 925 to 975 psig with tFa system vented outboard of the' check valve.
Acceptance criteria are to be determined by our Nuclear Engineering Department to be consistent with respect to the mission of the check valve (to prevent reverse flow) and the downstream overpressure protection. The test described in this response will be performed during each refueling outage.
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Item-C.I.c in Attachment 4 to your letter NRC Question:
~ g.
indicates that a check valve position indication system will be designed. Will this indication system supplant more quantitative means of verifying valve position (e.g., leak rate testing)?
BECo. Response: The statement in our letter and in the RHR Task Force report is that the check valve design providing position indication would be based on an evaluation of need. Any decision on replacement or modification of the check valve to install i
position indication will be made after designs are studied and the benefits determined.
Position indication devices, if added, will not replace the once-per-refueling outage leak test.
NRC Question:
h.
Will the RHR pressure gauges be used on a routine basis to check system pressure? If -
so, at what frequency will they be checked?
Will they be located in both RHR loops?.Will they be alarmed? Wil,1 they be used during valve operability testing to ensure valve closure? What will be their calibration frequency? Where will they be read out?
What will be lost by the removal of the.
pressure gauge in item C.3.c?
l BECo Response:
RHR's' system pressure gauges will be used on an as needed or on demand basis, i.e'. upon aldrm on RHR system high pressure or upon discovery of increased pipe wall / component temperature on the system.
Pressure gauges will be installed on both RHR loops.
The existing alarms from pressure switches pS 1001-74 A and B will still be used. The pressure gauges will normally be isolated and valved out of service.
Calibration of the pressure gauges will be once per refueling outage and upon request. The gauges in between the MO 1001-20 and MO 1001-29 valves will read out locally in the vicinity of the respective equipment rooms. The gauges outboard of the MO 1001-28 valves will read out locally in each RHR quadrant near the system instrument rack. The gauge removed by recommendation C.3.c will be replaced by the gauge between MO 1001-28B and 29B and the local vent for the injection line will be gained.
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NRC Question:
1.
What sp3cific cetion will be takan if th? RHR-systcm high pressure alare annuncictes in ths control room? A copy of the revised alarm procedure (item C.S.a) should be submitted to Region 1 prior.to restart. What are the maintenance and calibration histories of the RHR high pressure alarm switches?
BECo Resporse:
Specific action to be taken per the revised alarm procedure 2.3.2.1 (Attachment 3) is to log the occurrence of the alarm and to let down pressure as directed by the alarm procedure.
'If' the alarm occurs at a frequency greater than once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, then action will be initiated to investigate ard determine the leak rate.
If a leak rate 1 gpm is exceeded, a unit shutdown will commence. A separate procedure, TP B6-85 (Attachment 4) has been prepared to accomplish the investigation / evaluation of the RHR leakage with respect to the controlled leakoff. After trial use as a temporary procedure, the method will be incorporated as part of the RHR system operating procedure.
The maintenance and calibration histories reveal that pressure switches pS 1001-74A and B have not required maintenance since initial plant startup (1972) and are not on a periodic calibration schedule. The actuation point of the switches as found was conservatively set low (approximately 360 psig as found versus 392 psig set point). The pressure switches will be placed on a once per cycle calibration schedule.
NRC Question:
j.
How will the RHR system be vented? Will the "A" RHR loop also require venting between the 28 and 68 valves?
BECo Response:
The RHR system will be vented from four local high points.
One vent ic located on each loop of the tPCI injection lines and one on each' containment spray line. This venting will be performed on both loops per TP 86-84 (Attachment 5).
Venting on the LPCI injection lines will be performed between the MO 1001-29 and 28 valves when the reactor is at pressure.
NRC Question:
k.
How will RHR system temperature monitoring be conducted and at what frequency? What locations will be monitored (relative to injection valves and other RHR components)?
BECo Response:
Temperature monitoring devices in the form of adhesive temperature sensitive markers will be placed on the valve bodies / bonnets end uninsulated metal surfaces in the RHR system from the containment wall to the NO 1001-280 and A valves. There is approximately 9" of uninsulated piping Al-5 L
L_____.__.
outboard. of' the NO 1001-28B valve which will h ve'o markar.
.The marke.rs will be chacked once par weak for changa,.If a y.1 chInga in temparatur,a, is datected, or th RHR high prossure L
-alarm (from PS 1001-74B or 74A) annunciated a portable j;
temperature monitoring device will be used to measure local L
pipe wall temperature.
This is part of the investigation 1~
procedure (1P 86-85: Attachment 4)'.
i_
NRC Question:
1.
What is the schedule for completion of items C. I.c, C.4.a. and C.4.b?. With. regard to the footnote in Attachment 4 to your letter, long term action items should be included in the next Long Term Plan revision submitted to-NRR.
However, the proposed schedule should also be submitted to Region 1 prior to-restart of the plant.
BECo-Response:
Proposed schedules for RHR Action items are:
1.
Item C.1.c " Evaluate the feasibility of replacing or redesigning check valves to provide positive position indication." (This.is a long term item and should be based on a needs evaluation.)
g
.BECo will perform a review of industry experience with this application and a design review with GE, Bechtel, and the valve manufacturer (Rockwell). This design review started 5/15/06 and is expected to be completed by approximately 12/1/86. Any design modification recommendations which may result.from this review will then be scheduled via the Long Term Plan.
2.
3 tem C.4.a "IS1 program opens both 60A and 688 check valves once per quarter, if in cold shutdown change frequency to once per refuel.
This action is long term because the IST program relief may not be possible."
This recommendation has been reviewed by the coordinator of the Inservice Test (IST) program. The response to this proposal is that a lengthening of the test freauency for valve opening is not recommended. The ASME code Section XI of 1980 edition (Winter,1980 addenda) and 10CFR$0.55a(g) both indicate that since the RHR injection check valves (valves 2001-60A and B) are Category C valves and that a frequency of quarterly is appropriate. The test frequency of once per quarter if in cold shutdown is permitted by an approved relief request to our TST program. An increase of test frequency cannot be said to be impractical per 10CFR50.55a(g)(4) since it has been demonstrated that the test is practical as performed at its present frequency.
Therefore, it is concluded that this recommendation will not be implemented.
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1001-28B and 29B were 3.
Item C.4'.b " Reduce frequency of test.
stroked 62 times betwean 11/84 and 2/86. This acticn is long
- term beccuse it requires extensive loch Spec chinges." '
The proposed changes to Technical Specification would include all ECCS system-testing; the major change would be on the subject of test frequency.
In simplified terms, Technical Specification requirements to test. redundant components upon entry into an LCO will be. replaced with a philosophy that if the redundant-component is within specifications and its l
surveillance frequency is current, then additional testing would not be required.
An anticipated schedule for these changes is that submittal would take place (9) months from I
now (or March, 1987). This activity will be included in our l'
Long Term Plan.
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, CRC Quastion:
Discuss tha dasign review cnd implementation.
process associated with the 1983 modification of the MSIV's.
-BEco Response:
Boston Edison recognized its responsibility for the design adequacy of the Main Steam Isolation Valve. redesign and' followed its standard design change process in designing and'
. implementing the 1983 modification. The design change objectives were achieved by the~ redesign and with the-setscrew problem ~ resolved, the. valves will perform better than the original design.
l Boston Edison's standard design change process includes a number of controls to. ensure the adequacy of the final product. The process begins with a requirements analysis to-identify the main objectives of the change. This is followed L
by a scope, justification and approval phase where I
organizational agreement is reached relative to the l
modification.
Next, engineering resources prepare a conceptual design package (if applicable) which includes more l'
specific information regarding the pl,anned modification.
A multi-disciplined review of the conceptual design is performed including a constructability review. Upon receipt of organizational approval uf the conceptual design, the engineering department proceeds with detail design. The detail design results in a plant Design Change package j'
(PDC). This package receives a multi-disciplined review, design review board review and approval, constructability lc review and then is forwarded to the station for review.
At the' station the package is routed for review to ORC members and once the package has been approved by' those individuals, it is released to the Station Manager for implementation.
In this specific case, the design process was initiated in r
October of 1982, well in advance of refueling outage'six.
The requirements analysis, ' conducted over a four month period through January 1983, consisted of a review of plant-specific and industry-wide operating experience with the MSI5/'s.
i, Industry problems related to valve leakage reported in the US NRC Inspection and Enforcement Branch's Information Notice 82-23 were to be addressed by the design change as well as plant-specific problems related to main valve poppet and guide wear and valve stem failures (Licensee Event Reports78-019 ard 82-36).
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[we'rbgg[stIEVy mainisnhn'o[enhinhoring, thi volvTo ~
c manufceturer cnd th2 G2neral Electric Boiling Water Rocctor.
. Owner's Group MSIV Subcommittoo which was supported by th?
y Instituto of Nuclear Power Op2 rations; Electric Power Research Institute and Nuclear Safety Analysis Center, The requirements analysis phase. culminated in the selection by engineering maintenance and the valve manufacturer of an optimum set of valve improvements, A project scoping document justifying the set of improvements was prepared by engineering and maintenance, issued by both the Engineering and Operations Managers and was approved by the Vice President.of Nuclear Operations. The scope agreed upon included provisions.to' resolve the. problems identified in the requirements analysis phase.
MSIV leak-tightness was to be improved by a revision to the main poppet resulting in an elongated configuration for better seating of the main poppet. A self-aligning pilot poppet design was to be provided for improved pilot poppet seating. Guide and poppet wear was to be reduced through the use of a poppet anti-rotation design feature. Stem failures l
would be eliminated by increasing the stem diameter, eliminating a stress riser on the stes. at the backseat area and by using the floeting pilot poppet.
The detail design change process began in June of 1983_ and followed standard practices and procedures. The valve l
manufacturer was contracted to design and manufacture parts
)
to modify the MSIV's, Deliverables procured included-L assembly drawing, parts, installation instructions, machining drawings for modifying existing parts, instruction manual and design report. The Nuclear Engineering ~ Department received and approved the valve manufacturer's design' information and used it in preparing-the PDC.
Following design review board review of the PDC, a pre-implementation meeting was held for the MSIV.
modification. The-sole purpose of the meeting was to generate complete, clear and effective work instructions to implement the modification. The framework used was'the manufacturer's installation instructions which had already been reviewed and approved by engineering.
Participants included personnel from maintenance, design engineering, the valve manufacturer, quality control, field engineering and field craft supervision. The work instructions were agreed upon and the final assembly procedure was added to the PDC.
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_This sat of documents specified th> ass:mbly steps to b2
- follow:d, callsd out hold cnd witn2:s points cnd provid:d for sign-off by quality control, field engineering, cnd i
construction supervision. A Maintenance Request provides control of the activity and is required for any work at the i -
- station, Implementation of the MSIV modification was governed by a Maintenance Request. The Maintenance Request approval sequence includes a statement of authorized work scope, review by Quality Control (and insertion of their
~
quality requirements) preparation of plant systems for performance and the Watch Engineer's approval to start work.
After the PDC was issued for construction and the MR approved, Boston Edison implementation was assigned to the maintenance section with support to be provided by the site l
Resident Contractor. Then the Resident Contractor prepared detailed assembly documents called Work Process and Inspection Reports (WPIR).
The package was implemented in accordance with our standard work processes and procedures. Upon completion of the physical construction work, the documentation was submitted by the Resident Contractor to maintenance for post-work testing and final acceptance.
Mode' Switch NRC Question:
Please submit a written assessment of the loose wires and drawing / wiring discrepancies that were identified during your investigation. The assessment should address what has found, the implications with regard to the unanticipated primary containment isolations, other safety implications, and your corrective actions. The assessment should provide a basis for concluding that similar conditions do not exist in other safety related systems.
BECo Response:
A review' of loose wiring found during the PCIS system walkdown showed that it was possible to come up with combinations that could lead to an inadvertent PCIS initiation. The evaluation team concluded that, although it is unlikely that these comFI,'ations would have repeated themselves during two successive. shutdowns, at approximately the same' time during the shutdown process, they could not be discounted as potential causes to the events of 4/4 and 4/12.
It should be noted that terminal board terminations listed as being loose, were not excessively loose; a quarter of a turn on the terminal screw tightened the lugs to the acceptable point. The loose neutral terminations were felt to be more of a contributor than the terminal board connections.
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1 The loose neutral terminations were found to be a result of improper compression lug size on the PCIS, RPS neutral busses. The existing lugs were meant for wire sizes larger than #12AWG, as is used in C915, C917, and C916.
In mast cases, only one wire was connected under each lug adding to the possibility of bad terminations
.If a number of wires l
were landed under each lug, the problem may not have existed.
In some cases, the wires were found to be improperly stripped; this problem was found to have existed f rom the day the plant was built and there probably would have been no reason to check these terminals until now.
l These loose terminations have been corrected using acceptable compression lugs.
An investigation is being performed by the i
Nuclear Engineering Department to assess the safety significance of the findings.
A thorough walkdown of similar design safety related panels i
indicates that the use of incorrectly sized compression lugs are limited to the PCIS/RPS cabinets C915, C916, and C917.
Therefore, no other safety systems are affected by this problem. All circuits identified as having loose terminations would have failed in the safe direction causing either a partial or full pCIS or scram condition. Therefore, et no time was the safety of the publ,1c or the safe operation of the plant compromised.
A review of the wiring discrepancies, noted during the investigative walkdowns, indicates two types of deviations, the first being typographical errot s on the associated prints, and the second being electrically equivalent circuits. In no case, was a circuit found to not operate as designed or as intended by the elementary control diagrams.
A contact off the high water level switch in the bypass circuit of the Main Steam line low pressure PCIS trip on Drawing MIN 33-10 was the only exception. This error is an omission from the elementary diagram and has since been corrected. The bypass circuit is wired and operates as designed and is shown on all other associated electrical prints.
The team concluded that there is no reason to believe improper safety system operation could result from drawing discrepancies because of the above findings. The MIN 33-10 discrepancy is being treated as an isolated case where it was captured on all other associated prints and no similar discrepancies were found.
Print discrepancies have been noted and forwarded to the Nuclear Engineering Department for corrections on 5/18/86.
Current quality control practices should eliminate the recurrence of the f.mproper use of compression lugs in any plant wiring in the future.
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On) troubleshooting'techniqu3 urcd by the code switch tsem
~'
was a " hind over hend" wiro ch2ck. This involved physically-cnd visunlly trccing wires from point to point in thi suspect E
logic circuits. This method was employed to assure tha term that the panels were indeed wired as shown and no new or extra wires had been added to the logic circuit incorrectly.
This check revealed'several circuit problems.
6 The following is a brief listing of the logic sub channels (pCIS or RPS) affected and a brief description of the problem-
-found.
As.can be seen by matching.the subchannel comb.47tions, it is possible to come up with combinations that cus:d cause a full pCIS or a full reactor scram. The fact that this is possible, does not make it probable that two wires wou4d'become loose at exactly the same time in two different controlled reactor shutdowns eight days apart.
PCIS Subchannel Affected l.
Al Loose fuse - affected 16A-K7A contributed to a Group I isolation channel (MR-86-45-189 was issued to correct).g 2.
Al Terminal block loose screw connection panel C2205A affects Yarway hi water level isolation signal - directly affects 16A-K7A (Group I isolation) when not in "RUN" mode (MR-86-45-193 was issued to correct).
3..
A2 Panel C905 terminal block loose screw - directly affects main steam line low pressure bypass with the mode switch in "RUN" - ec
- i cause a 1/4 Group I isolation signal channel A2 by de-energizing relay 5A-K7C (MR-86-45-190 was issued to correct).
4.
B1 Lnose wire at RK2205A - affects bypass around main condenser low vacuum and main steam line valve closure - could cause 1/2 scram from subchannel B1 (MR-86-45-192 was issued to correct).
5.
B2 Loose screw connection at RK2256B - affects relay 16A-K3D (steam line hi flow) - directly affects 16A -K7D (Group I isolation) (MR-86-45-191 was issued to correct).
6.
Scratu Reset A and Channel A1 and A2 Neutral bus wire out of lug and resting on neutral bus - affects relays 5A-K18A&C, 5A-K19A&C and relay 5A-k22A - could cause auto scram channel Al and A2 due to not bypassing scram discharge high water level - would not allow subchannel Al and A2 to be reset (MR-86-300 was issued to correct).
Al-12
RPS Chann,1 7.
B1 Neutral wir0 loose under compression lug offects relays 5A-K3B, 5A-K3f, SA-K4B and 5A and SB - could cause subchannel B1 1/2 scram if reactor pressure was less than 600 psi also alarm typer alarm " Main Steam Isolatiors Valves Not Fully Open Scram Trip" for channel B (MR-86-303 was issued to correct).
8.
Al Neutral wire completely out of compression lug - affects relays 5A-K8A, 5A-K9A, 5A-K10A and 5A-K10L - could cause 1/2 scram subchannel Al and annunciators " control valve fast closure scram trip," " generator load rejection - turbine stop valve scram bypass "
turbine stop valve not fully open scram if first stage pressure greater than 45% (MR-86-302 was issued to correct).
9.
Al Panel C915 - Neutral bus terminal block EJ point #9 found loose wire under compression lug - affects relays 5A-K3 A, 5A-K3E, 5A-K4A and SA-KSA - could cause an 1/2 auto scram channel A1; annunciators "MSIV not fully open scram at RX pressure > 600 PSI", "Drywell high pressure scram trip", and " Reactor vessel high pressure scram trip";
and computer alarms C1509
" Main steam line isolation valves not fully open scram trip", c1513 "Drywell high pressure scram trip",
and C1517
" Reactor vessel high pressure ser,am trip"(MR 300 was issued to correct).
- 10. Al Panel C915 - Neutral bus terminal block EJ point #10 found loose wire under compression lug - affects relays 5A-k2A, 5A-K11A, and SA-K1A -
could cause an 1/2 auto scram channel A1; annunciators " Main condenser low vacuum scram trip", " Discharge volume high water level CRD scram trip", and " Condenser low vacuum and main steam Isolation valve closure scram bypass"; and computer alarms C1505
" Condenser low vacuum scram trip", and C1501
" Discharge volume high water level scram trip"(MR-86-300 was issued to correct).
- 11. A2 panel C915 - Neutral bus terminal block BJ point #1 found loose wire under compression log - affects relays 16A-K4C, 16A-K3C, 16A-K2C, and 16A-K1C - could cause an 1/2 Group I isolation from PCIS channel A2; annunciators " Main steam line low pressure", " Main steam line Hi flow", " Steam tunnel Hi Temp." and " Reactor vessel low low water level"; and computer alarms C1543
" Main steam line High Flow" and C1547
" Steam tunnel High Temp."(MR-86-300 was issued to correct).
- 12. A2 Panel C915 - Neutral bus terminal block BU point #2 found loose wire under compression lug - affects relays 164-K44C, 16A-K7C, 16A-K6C, 16A-KSC, 16A-K5BC, 16 A-K59C, and 16 A-K19C.
This could cause an 1/2 Group I isolation from PCIS channel AZ; 1/2 isolation of T.I.P.
withdrawal, RHR shutdown cooling, head spray and discharge to Radwaste valves; 1/2 isolation of RWCU system; a trip signal to the reactor building isolation and standby gas treatment initiation
- ystems; annunciators " Exhaust vent high Rad", and " Vent exhaust monitor down scale"; and computer alarm C1168
" Refueling floor vent exhaust Rad"(MR-86-300 was issued to correct).
A1-13
)
- 13. A2 Panel C915 - Neutral bus terminal block BJ point #9 found loose wire under compression lug - cf fects roleys SA-K10C, 5A-K10G, 5A-KBC, cnd SA-K9C, This could cause an' 1/2 auto scram channel A2; annunciators
" Turbine stop valve not fully open scram if first stage pressure >
45%", " Control valve fast closure scram trip", and " Generator load rejection & turbine stop valve scram bypass"; and computer alarms C1555
" Turbine stop valve closure scram trip", and C1559
" Control valta fast closure scram trip"(MR-86-300 was issued to correct).
- 14. A2 panel C915 - Neutral bus terminal block BJ point #10 found loose wire under compression lug - affects relays 5A-K2C, 5A-K11C, and SA-K1C.
This could cause an 1/2 auto scram channel A2; annunciators
" Main condenser low vacuum scram trip", " Condenser low vacuum & main steam isolation valve closure scram bypass", and " Discharge volume high water level CRD scram trip"; and computer alarms C1503 = " Discharge volume high water level scram trip", and C1507
" Condenser low vacuum scram trip"(MR 86-300 was issued to correct).
- 15. A2 panel C915 - Neutral bus terminal block BJ point #12 found loose wire under compression lug - affects relays and contactors 5A-K14C, 5A-K14G,. 5A-K19 A, and 5A-K19C.
This could cause an 1/2 auto scram channel A2; annunciator
" Reactor auto-scram channel "A"; And computer point C1539
" Reactor auto scram chpnnel "A"(MR 300 was issued to correct).
- 16. A3 panel C91b - Neutral bus termine) block CS point #9 found loose wire under compression lug - affects relays and contactors 5A-K16A, 5A-K17A, SA-Kt%A, 5A-K15C, o A-K19C, and 5A-4 3 9 A.
This could cause an 1/2 manual scram channel A3; annunciators
" Mode Switch shutdown scram bypass", and " Reactor manual scram channel "A"; and computer alarm C1537
" Reactor Manual scram channel "A"(MR 86-300 was issuod to correct).
- 17. B1 panel C917 - Neutral bus terminal block EJ point #6 found loose wire under compression lug - affects relays SA-K12f, 5A-K120, Sa-K70, 5A-K6B, 5A-K278, and 5A-K2BB.
This could cause an 1/2 auto scram channel B1; annunciators
" Reactor neutron monitoring system scram trip", " Main steam line high radiation scram trip", " reactor vessel low level scram trip", and " Discharge volume high water level CRD scram trip", and computer alarms C1530
" Neutron monitoring system scram trip", C1526
" Main steam line high radiation scram trip",
C1522
" Reactor vessel low water level scram trip", and C1586 -
" Discharge volume high water level scram trip"(MR-86-301 was issued to correct).
- 18. B1 panel C917 - Neutral bus terminal block EJ point #10 found loose wire under compression lug - af fects relays 5A-K28, SA-K11B, and SA-K18.
This could cause an 1/2 auto scram channel 01; annunciators
- " Main condenser low vacuum scram trip", " Condenser low vacuum &
main steam isolation valve closure scram bypass", and " Discharge volume high water level CRD scram trip"; and computer alarms C1506 -
" Condenser low vacuum scram trip", and C1502
" Discharge volume high water level scram trip"(MR-86-301 was issued to correct).
Al-14
19.'B2 Penel C917 '- Neutral bus terminal block BJ point #1 found loose wire l
L und2r compression lug - Offects relays 16A-K60, 16A-K50, 160-K58D, 16A-K590, and 16A-K19D. This could.cause an 1/2 Group I isolation from PCIS channel G2; and 1/2 isolation of T.I.P. withdrawal,. RHR shutdown cooling, head spray and discharge into radwaste valves 1/2 isolation of RWCU system; a trip signal to the reactor building isolation and standby gas treatment initiation systems; annunciators
- " Exhaust vent high Rad", and " Vent exhaust monitor downscale"; and computer alarm C1168
" Refueling floor vent exhaust Rad"(MR-86-301 was issued to correct).
- 20. 83 Panel C917 - Neutral bus terminal block CS point #9 found loose wire I
under compression lug - affects relays and contactors 5A-K168, SA-K17B, SA-KISB, SA-KISD, SA-K19D, and. 5A-K198. This could cause an 1/2 manual Scram channel B3; annunciators
" Mode Switch shutdown I
scram bypass", and " Reactor Manual scram channel. 'B'"; and. computer alarm C1538
" Reactor manual scram channel 'B'"(MR-86-301 was issued to correct).
L
- 21. B1 Panel C917 - Neutral bus terminal block CS point #12 found loose wire under compression lug - affects relay SA-X13B. No effect with RPS non-coincident neutron monitoring shorting links installed (MR-86-301 was issued to correct).
Wiring & Print Discrepancies 1.
Print MIN 33-10 (Rev. EO) does not show 16A-K19A contact 1-2 in the mode switch bypass circuit.
Error identified on PCAQ" to Engineering; panel wiring diagram shows circuit as designed.
Schematic to be revised to reflect change.
2, Print MIP 426-16 (Rev. ES) shows E311 going to 07-14 (SA-K3A) but in fact goes to DY-14 (5A-K19A).
See 3tems 3 and 15 (Electrical Equivalent).
PCAQ written to update print,
3.
Print MIP 426-16 (Rev. ES) shows EJ9 going to DQ-14 (5A-K10A) but in fact goes to DZ-14 (5A-K3A).
See Items 2 and 15 (Electrical Equivalent). PCAQ written to update print.
4.
Print MIP 426-16 (Rev. ES) shows wire designation on DH-7 (16A-K7A) as DD-8 and it should be DD-18.
Typo on print.
PCAQ written to update print.
PCAQ IS A " Potential Condition Adverse to Quality" report which is used to effect review and resolution of an identified condition which is potentially adverse to quality.
Al-15
( bi Print M3P 421-16 (Rev. E5) shows a-wire on AF-14 (16A-K4D) going to 832 L
- but in, fact-is not thare. -Also BJ2 shows wire going to AF-14 (16A-K4D) but'is not thtre.
O
- Actual wiring is Electrical Equivalent. PCAQ written to update print or-
. change wiring to agree with print.
6.
Print M3P 423-16 (Rev. ES) shows.only one wire on AK-14 from A3-14.
(16A-K70) but in fact has a second wire on it going to DG-14 (16A-K44B).
Similar to Item $.
.This wire causes Electrical Equivalent circuit.
PCAQ written to update print.
1 I'
7.
Print MIP 423-16 (Rev. E5) shows only one wire on DG-14 (16A-K44D) from DH-14 (16A-K70) but has a second wire on it going to AK-14 (16A-K440),
Same as Item 6 but in opposite channel.
j PCAQ written to update print.
l 8.-
Print MIP 423-16 (Rev E5) shows terminal block point 7 going to DB-3.
It should read DB-13.
Typo error on print.
PCAQ written to update typo.
Print MIP 423-16 and lamicoid marker in Panel C917' show fuse 16A-F64A 9.
going to C917 terminal block DD point 42.
Print' MIN 34-9 shows fuse 16A-F64B going to C917 termins.1 block DD point 42.
MIP 426-16, MIN 33-10 and lamicoid marker in C915 show 16A-F64A going to DD point 42 in Panel l
915.
Drafting error on print MIP 423-16. F28(16A-F64A) should read 16A-F640.
l.
PCAQ written to update print.
- 10. Print MIP 421-16 shows two F29's on terminal block BB - one should be F28 which is also designated as 5A-F270.
F29 is designated 5A-F28D.
Typo error on print MIP 421-16.
PCAQ written to update typo.
- 11. Print MIP 460-8 (Rev. E4) shows only one wire on device AN terminal 2 going-to terminal block BB point.97. In fact, there is a second wire on device AN terminal 2 going to device AY terminal 9.
Print shows AY-9 going to 1B BB97 and TB BB97 to AY-9.
This wire does not exist.
Wiring is per elementary diagram.
This is a drafting error on MIP 460-8.
PCAQ written to update print or change wiring to agree with the print.
- 12. Frint MIP 460-8 shows device BA terminal 10 going to device 80 terminal 14 but this wire does not exist. Device BO terminal 14 has one wire on it 9oing to device BP terminal B4.
Device BP terminal B4 has two wires on it but print only shows one wire, the other wire goes to devir2 BO terminal 14.
Actual wiring is an Electrical Equivalent.
PCAQ written to update print or change wiring to agree with the print.
Al-16 l
l Ja_________.___-__--__-
13.' Print MIP 459-9 (Rev'. [4)~ device BF terminal.10 has two wires on it going to torminal block CC point 89.
Print cnly shows one wire.
C This is a_ duplication of wiring.
PCAQ written to remove wire or leave as is.
14 Print MIP 426 *.6 (Rev. ES) shows device DG terminal 13 going to' device EG terminal 2; it should be terminal 12.
Print MIP 426-16 (Rev. ES) shows.
. device DH w2th two wires on terminal 3, the second wire going to device DC terminal 2 should be terminal 13.
This.is a typo error on 9IP 426-16.
PCAQ written to update print.
- 15. Print MIP 426-16 shows neutral bus EJ point 8 ~ going to device DY terminal 14.
EJ-0 goes t.) device DQ terminal 14.
See Items ' 2 and 3.
Electrical Equivalent PCAQ written to update print ~.
- 16. Print MIP 460-8 (Rev. ES) device BF terminal 10 has two wires on it going to terminal block CC point 89.
Print only shows one wire.
Similar to Item 13.
PCAQ written to change wiring or leave as is.
- 17. Print MIN 27-16 (Rev. E5) relay tabulation does not show contacts 9-10 and 11-12 of relay 16A-K13 being used.
In fact print MIN 36-7 (Rev. E6) does show them being used. Contact 9-10 is in 16A-D5159A circuit and contact 11-12 is in reset-circuit.
Contact usage was emitted from print MIN-27-16.
Actual conditions were per print MIN 36-7.
PCAQ written to update MIN 27-16.
It should show it going to DD-FS-1.
Typo error on the print.
PCAQ written to correct MIP 426-16.
A1-17
_ TASK L
- Based on the miswiring found'in the Reactor Mode Switch related circuitry (Reactor Manual Control System):
1 1.
Evaluate the'effect(s) on MSIV closure.
2.
Determine if this discrepancy was more or less conservative than the cesign prints.
REFERENCES MIV Series Drawings; Elementary Diagram Reactor Manual Control System PNPS FSAR, Rev. 5.
DISCUSSION OF MISW!R1WG As a result of the performance of step 8 on page 11 of PNPS Procedure TP 86-64, Remove Existing SB-1 Type RMS, Replace with SB-9 Type RMS, a discrepancy was found between the Elementary Diagram MIV16-6 wiring and the "as found" wiring. _ The dif ferer.ce is shown below:
MIV16-6 (Should Be)
As found 00 73 go.q g ~
i.
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> A' *M g A.wu g A.sn m.m
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configuration,1 relay 3A-K21 can be eaergiged independent o7 Rke postQ1on o7_
RMS contact 30-30C and is solely dependent on the position of RMS contact 29-29C,')
e., the contacts are in parallel, while in-the "as found" configuration, relay 3A-K21 is tependent on the. position of both RMS contacts29-29C and 30-30C,-i.e., the contacts are in series. As a result, 3A-K21 could never be energized with contact 30-30C open, which is in conflict with the design' diagram.
This concept is outlined in the following table where blank.open, X closed. 0 de-energized, i. energized:
POSSIBLE RESULTANT RESULTANT CONTACT "SHOULD BE" "AS FOUND" POSITION RELAY RELAY COMBINATIONS CONDITION CONDITION 29-29C 30-30C 3A-K21
-3A-K23A,B 3A-K22 3A-K21 3A-K23A,B 3A-K22 0
0 0
0 0
0 X
0 0
0 0
0 0
X 1
1 0
0 0
0 X
X 1
1*-
1*
I
'1*
l'
The following table shows the relationship between the RMS position and the 29-29C and 30-30C contact positions (taken from Elementary Diagram MIN 14-4) t.ere blank.open, X-closed:
RMS S/U &
CONTACT RUN INTER S/B INTER REFUEL INTER S/0 X
29-29C 30-30C X
X X
X l
As can be seen from this table, the positioning of the mode switch never requires contact 29-29C to be closed with contact 30-30C open.
The only position in which contact 29-29C is required to be closec is the " Refuel" position in which contact 30-30C is coincidentally closed.
PURPOSE OF CIRCUIT l
The circuit which was effected by the miswiring is part of the Refueling l
Interlock portion of the Reactor Manual Control System. As outlined in l
Section 7.6 of the FSAR.. Refueling Interlocks:
Safety Design Basis During fuel movements in or over the reactor core, all control rods 1.
l.
shall be in their fully inserted positions.
No more than one control rod shall be withdrawn from its fully 2.
inserted position at any time when the reactor is in the refuel mode.
Al-19 1
]
The refueling interlocks. reinforce operational. procedures that prohibit taking
~
the. reactor critical under certain situations encountered during refueling operations by restricting the movement of control rods and operation of The effected circuit is designed to allow movement-of
., refueling equipment.one control rod with the mode switch in the " Refuel" position and rods fully inserted provided all other interlocks are satisfied.
RESULTS Using the relay tabulation on Elementary Diagram MIV8-5, all relays and contacts in the effectec circuit were traced through drawings for their potential resultant action / annunciation.
The results of the analysis show No the only interfaces to be with refuel and rod block interlocks.
annunciations or RPIS computer printouts could have resulted from the miswiring.
ANALYSIS OF POTENTIAL FAILURE MODES Because of the nature of the miswiring found, the only failure mode of corcern would be when the mode switch is in the " Refuel" position.
If contact 29-29C were to fail open in the refuel position while contact 30-300 were to close, the effect on the circuit woufd not be altered by the In either condition contact 3A-K21 would remain miswiring that was found.
de-energized, and as a result, single rod withdrawal in the refuel mode could not be performed.
If contact 30-30C were to fail open in the refuel position while contact 29-29C were to close, relay 3A-K21 would again remain de-energized in the "as As a found" circuit while it would have energized in the "should be" circuit.
result, single rod withdrawal in the refuel mode would not have been able to be performed as desired.
CONCLUSIONS In response to Task 1, analysis of the Reactor Mode Switch circuitry and the Reactor Manual Control System Circuitry shows no potential effect on the MSIV The RMS miswiring can in no way cause or prevent an MSIV closure.
closure.
In response to Task 2, since the miswiring could only result in the inability l
to withdraw a rod when desired, the resultant condition was more-conservative than the design prints.
AI-20
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-NRC Qu2stion:
How ere chtnnel sep3rstion of Q cnd non-Q circuits j
cnsured b3twesn th2 mor.itored points cnd th2 GETARS '
system?
~h BECo Response:
Channel separation between the GETARs Monitoring System and the actual monitored points are insured in j
I three ways:
1.
The points monitored with the use of dry contacts are separated from the FCIS logic by the relays themselves.
I The G.E. Century 100 relays provide _ mechanical and electrical isolation as a design function of the relay.
All inter-panel wiring for C915 and C917 will be Class 1E. 'The GETARs System is fused at 1 amp 120V, and since the Century 100 relay contacts are rated at 12 amps continuous, there is no possibility of welding contacts closed, preventing the relay from opening.
2.
Voltage divider circuits are used in the channels monitoring the actual PCIS logic channel.
A 2 megohm resistor is used to provide the isolation.
Any failure of the monitoring system could not affect the PCIS logic due to this high resistance valup.
The 2 megohm resistors were bought to Q requirements and are wired class IE.
3.
Both A and B logic systems are separated from each other by separate input multiplexors.
These multiplexer are separated from the main GETARs Computer by fiber optics cable, This provides added insurance that a GETARs failure would not affect both logic channels.
In summary, the GET ARs Monitoring System is separated from Q circuits by the physical constructions of the relays themselves and by the Q resistor voltage divider circuits used in the logic monitoring channels.
As added insurance, the A and B pCIS logic channels are separated from each other by separate input multiplexer.
Al-23 L
Q.,
,4-4 TABLE Recommendation # '
Summary I
B.1' Mitigating the effect'of (RHR) leakage by allowing a controlled amount of leakage to exit the system.
C.I.b Perform once per/ refueling outage leak test on 6B's for pressure drop capability (to be
. initiated in RFO #7).
C.1.c Evaluate the feasibility of replacing or redesigning check _ val,ves to provide positive position indication.
C.2.a Install additional pressure gauges in system.
C.3.a Develop system venting program.
C.3.b Provide means for system temperature monitoring.
C.4.a Change frequency of 68A and 68B cheek valve testing from once per quarter to once per refueling outage.
F C.4.b
. Reduce' frequency of 28 & 29 isolation valve
' (stroke) testing.
C.S.a Revise alarm response procedure to allow for control of repressuritation and provide additional assessment and corrective action.
C.6 Trend surveillance history of 400 psig valve interlock for reliabiliy.
l' A1-24 o_
_ = _ _ _ _ _ _ _ _ _ _ _ _,
________________________________________________________._o
r."/
O i
ATTACHMENT 2 s
BECo Responses to NRC Questions As Presented at May 19, 1006 Meeting at Region 1 Office r
l 2
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I l
l i
I l
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detailedResultsofthe288ValveDis' assembly & Inspection 1,
Inspection results of valve MO 1001-288:
The disassembly of the NO 1001-280 valvo showed no abnormal or significant wear on the mating surfaces of the valve disk or seat.
A minor irregularity or mark was found at the 11 o' clock position on the valve seat.
Also, the stem locknut to disk nub weld was cracked but remained structurally -intact. The weld was repaired as a part of. this refurbishment. The valve was lightly lapped and then blue-checked.
The results of the blue-check were excellent.
The valve was inspected by VT-3 examination methods Data sheet 86-10-15 documents the inspection and is included in this attachment. The inspection results record evidence of erosion above the valve seating surfaces on the body of the main disk. The erosion varied in depth from approximately 3/16" to 5/16" for the full circumference of the disk.
The Nuclear Engineering Department (NED) was requested to review and respond to the acceptability of this findire.
NED has responded that the root cause of the observed volumetric wear on s
the disk is attributed to cavitation (erosion) pitting followed by concentration cell corrosion attack.
Specifically, cavitation pitting occurs initially while throttling during extended shutdown cooling service.
Concentration cell corrosion attack in the resultant pits subsequently follows.
The failure of the locking nut tack welds was attributed to poor weld technique during manufacture.
The wear noted is acceptable by analysis for interim operation as has been evaluated by NED to ASML III NB-3000 criteria.
NED concludes that the remaining metal significantly exceeds the minimum acceptable disk thickness for postulated loads including design, emergency and faulted conditions.
The wear rate will not result in an unacceptable reduction in disk thickness prior to refueling outage #7 (January 1987). This disk (for MO 1001-28B) will be removed and evaluated or restored during RFO-7 to confirm the wear rate. Disk deterioration in the form of a gross metal removal is not considered related to seat leakage. The NO 1001-288 valve will not have a motor operator torque setting in excess of 3 5/8 (as it is presently set).
As a consequence of the findings on the MO 1001-28B valve, NED has concluded that similar wear may have occurred in the MO 1001-28A valve and recommends that the 1001-28A valve be disassembled and inspected at this time. This is to ascertain whether comparable disk deterioration has occurred.
The valve disassembly is in progress in response to the recommendation.
A2-1
-__-____-___ -______ _ - _ __ a
f 2,
Postwork test results of valve MO 1001-280:
, After reassembly, the NO 1001-288 valve was stroke tested, Local Leak Rate tested with air per Appendix J of 10CFR50, and hydrodynamically tested (water test).
The table below summarizes the results.
LtRT Results MO 1001-28D (45 psig air)
As left RF0 6 As found April 1986 postwork May 1986-i p
y 1.9 SLM 1.5 SLM-2.0 SLM Hydro dynamic Test Results 950 psig applied Inboard of Check Valve 1001-688 i
April 1986 May 1986 MO 1001-29B and MO 1001-29B and MO 1001-20D MO 1001-28B closed MO 1001-28B closed-closed
- 0. 3 3 g pm 0.48 gpm 0.38 gpm The post assembly air test results are 0.5 SLM higher than the April 1986 as-found results,.The disassembly and lapping of MO 1001-28B did not result in an improvement in leakage rate. ' Test methods used were the The April results may have been lower because the valve was last same.
closed in the warm condition prior to test. The post assembly high pressure water test was conducted after refill of the system, the system was vented and tested with both the MO 1001-290 and 288 valves closed.
This leakage rate was higher than the April as found condition; however the test method uses the number of strokes that a positive displacement pump makes to maintain pressure. The April result was based on receiving an average of 4.75 strokes per minute where the May-result was 7 strokes in a 1 minute period.
The flow to the volume is not measured by a flow measurement device. The differences between the 0.33 gpm of. April and the 0.48 gpm is accounted for by the method to count pump strokes (ie, one minute vs ten minutes) and the isolation of valve 1001-338,'(the manual l
ble.k valve that forms a test boundary) may not have been as tight in May as in April.
The test conducted with only the 28B as boundary showed a lower leakage rate than with both valves as a boundary.
l Based on the postwork test results, the controlled leakoff method discussed in the answer to NRC question (e), was selected a' ~ur course of action.
R2-2 l
1 l
Attachment A PILGRIM NUCLEAR' POWER STATION
+
Data Sheet #,8(n-fD-/5 VISUAL EXAMINATIO ND VT-4
/[
'I have read and understood PNPS procedure 1.3.39 /Al) 6-2/-66 Examiner's Initials Date Categoryk,/A System /0 - FN/2 ASME Sec XI Class 8
1sometric or Print 757 T-/0-48 Rev. A MR # 8d-/C-/5 Component Description h6 $WE/?)//kJ - MO-/(ky-7$18 RwP g Location ? } llf7. llA U M 0 Visual Exam:
Direct _2_ _ Remote Requires Requires Observed Conditions Supplem Further Yes No Accept Reject Eram Eval Cracks Y /,
Corrosion
/ d J
Loose Parts y
v V
Misalignments Mechanical Damage t/
/ NW NK Erosion Other (Describe) g (ethattachp Visual Aids Used Ft. A S.4 L i 6 4 7 Ht(RR_o/2, Comments: hfLE ERON Mi LbM ou V'ALVf_ hl $C - SEF AirAc m s net c g Fs(2.t%~l'G Mhthedi
~
Supplemental Examination Data Sheet # (if applicable)
Examiner {ll_ D D Level I Reviewed By:
k A Qf f)
Leve)
Date $5\\ -bb l
l Sr. Mechanical Engineer (151) or Designee d
A-Date Authorized InspectM 1
A2-3 c.
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lif nosi m tas m h
v VALVE m
STEM s
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% ERoslou VARIES FROM k"Tok4
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'M
{PTH DISK SEATING SURFACE' 3
A2-4
Description of Control Room Stoffing (40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> pinn)' end Detnils _
Rcenrding Plans to Comply with 10CFR$5.32(ej Due.to theLcurrent labor action by Boston' Edison's union employees, qualified exempt personnel have been assigned to fulfill the shif t nanning requirements established by our Technical Specifications..
A three section watch bill is in effect. 'It has bren structured such that individuals are h
scheduled to. work not more than 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> per week, The limit established by our station procedures is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week.
AllL personnel assigned as nuclear watch engineers, nuclear operations supervisors, and licensed reactor operators have been actively performing the functions of senior reactor operator as required by 10CFR55.31. This requirement has been met by independently standing watch in the. Control Room during power operations.
Additionally, each of the above individuals have-previously held a reactor operator's license.
All of the above individuals.
completed simulator requalification training during the period October to December, 1985. This training includes hands-on manipulation of the reactor controls.
Prior to start-up from the current outage, routine surveillance
. testing must be performed on the following safety related systems:
High Pressure Coolant Injection g
Reactor, Core Isolation Cooling Low Pressure Coolant Injection Emergency Diesel Generators
. Core Spray Neutron Monitoring Reactor. Manual Control System This testing, which requires hands-on manipulation of the system's controls switches, will be performed such that each individual assigned as a licens, d reactor operator receives hands-on refresher training with some of the above systems.
Personnel' assigned as unlicensed nuclear plant operators (ie, auxiliary operators) are receiving on the job training in plant. tours and operation of auxiliary equipment. This job-specific training will be documented and will be a function of the experience of the individual assigned as unlicensed operators.
A2-5 l
.=-
l-l' BECo's' Di positioning ' of NRC Information Notics on "Stt Screw" Problems Experienced Elsewhere in the Industry During 'our presentation at the Region 1 office on May 19, 1986, and in-subsequent discussions with your staff, you. asked us to review seven (7)
~
Inspection and Enforcement Information Notices (82-20, 83-70, 84-36, 84-53,
. 86-01, 86-09, and 86-29) regarding set screw issues to determine applicability-
- to the MSIV Plant Design Change (PDC 86-28). The review has been completed and it has been determined.that in all cases the issues raised in the information notices are either not applicable or properly addressed in PDC 86-28.
The results of our review have been. tabulated in the following Table.
k t
A2-6
g-.
DOCUMENT TOPIC-DISCUSSION.
DISPOSITION ILN-82 Check' valve Aloyco thsck valvos ILN does not includs problem' mounted in vertical.
a set screw problom.
orientation sustained.
Damage occurred to internal damage to disk stud directly.
stud serving as stop, Pacific valve also Note BECo does not sustained similar damage use Aloyco or Pacific and a material defect check valves in limited to manufacturers Pressure' Boundary product line.
Applications. Velan, Rockwell and A/D valves used, either have integral disk stand off or are tilting disk.
ILN-83-70 Vibration Threaded fasteners Applications were not induced valve (Yoke to Bonnett. Stem MSIVs. Similar
-failures Clamp) had backed out applications on PNPS in non-MS3V applic,ations.
utilize locking tabs to preclude this (note pilot / poppet set screw is staked vs. lock tab).
IEN-84-36 Loosening of Set screw on the locking Set screws were not locking nut on nut for the worm gear staked. BECo went Limitorque backed out.
through and staked Valve Operator selected set screws on all Q MOV's.
This IEN points out the need for staking as already considered in PDC 86-28.
IEN-84-53 Information Excessive Lock-tite Lock-tite not used in.
concerning the not wiped of a scram.
MS]V due to this use of Lock pilot solenoid valve concern plus degrade-acorn nut fouled tion induced by valve other radiation. Use of anaerobic adhesive /
Lock-tite is sealant controlled to nuclear grades dedicated for use by Engineering and utilized under controlled process whereby administrative controls preclude excessive application (Ref RA&P-84-346)
A2-7 1..
E
TCBLE b.
DOCOMENT TOPIC OISCUSIION DISPOSITION
[EN-86-01 Tallure of Main Generic failure of As per IEN-82-20, BECo feedwater check Pacific ~ check valve dise doesn't use Pacific chec valves causes loss nut retainers essentially valves in Pressure of FW system identical to IEN-82-20~
Boundary applications.
integrity and above Valves used have differe '
water hammer integral disc stand off.
damage.
tilting disc.
Failure h.
no bearing on MSIV task.
No apparent use or failu-of set screws.
IEN-86-09 Failure of check Leakage past a normally No set screws involved, valves subject to closed motor operated record of similar occur-low flow valve in Auxiliary rence at PNPS, no similar conditions Feedwater turbine steam applications, except supply line fatigue RCIC/HPCI exhaust cycled stop check valvo lines which have not disc & disc guide studs experienced this problem.
Not. applicabl.
IEN-86-29' Effects of Change to TS bypass limit No set screws involved, changing opera-switch settings for PNPS does not use tor switch 100-85-03 at Songs-3 TS bypass, not appli-settings resulted in unintended
. cable to MSIVs in par-changes in closed ticular or PNPS in position indications.
g ene ra!..
Valves did not stroke full closed.
l l-A2-8
.j l
1 E___-________________________________________________________
_____________________]
ATT ACHMENT 3 Revised PNPS Procedure 2.3.2)
Alarm Procedure Panel 903 Left i
l l
EDISON
~
NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure No. 2.3.2.1 PANEL PROCEDURES PANEL 903 LEFT List of Effective Pages 2.3.2.1-1 2.3.2.1-2 2.3.2.1-3 2.3.2.1-4 2.3.2.1-5 2.3.2.1-6 2.3.2.1-7 2.3.2.1-8 2.3.2.1-9 2.3.2.1-10 2.3.2.1-11 2.3.2.1-12 1
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- .y ATTACHMENT 4 Temporarv PNPS Procedure TP 86-85 RHR Intersystem Leakage Assessment And Controlled Leakoff
BairsoISON PILGRIM NUCLEAR POWER STATION RHR INTERSYSTEM LEAKAGE ASSESSMENT AND CONTROLLED LEAK 0FF TP 86-85 i
List of Effective Pages TP 86-85-1 TP 86-85-2 TP 86-85-3 TP 86-85-4 TP 86-85-5 Attachments TP 86-85A-1 TP 86-85A-2' TP 86-858-1 TP 86-85B-2
[kM A6hred Approved M1earsp'erationsManager Date GllCl9C i
6!/6[PP Expiration Date A4-1 TP 86-85-1 Rev.
9g,
I.
PURPOEE: -
i
'!b provide station personnel with a method of establishing a controlled j
leakoff of the Mm Discharge Piping to prevent increases in pressure i
due to leakage past the RER Injection Isolation Valves. his procedure l
say be used to establish a controlled leakoff when the RER system becoues pressurize $ from leakage past its isolation valves, or in i
advance of such leakage.
II. DI!KIJSSION:
Leakage past th'e RHR Injection Isolation Valves, principally the l'
MOV1001-28B and MOV1001-29B, have lead to repeated annunciation of the "RHR ' Discharge or Shutdown Cooling Suction High Pressure" alarm. Prior to maintenance on the MOV1001-28B valve, leakage was seasured at.33 gpm at 975 peig. his procedure will establish a controlled leakoff from the PHR Pung Discharge Piping to the 'rbrus via the 2" 1000 valve bypass around the Punp Discharge check valve. Eis will prevent PER puup discharge piping pressure from pressurizing wtrD in-leakage rates less than 1.0 gpm are present.
III. REFEPDCES:
A.
PsD M-241, Sheet 1 & 2 B.
RER Task Force Report, Revision 1 Dated 5/5/86 C.
ESR Response 86-153 Dated V2V86 D.
PDC 86-30 IV. PRE:RDODISI'IES:
A.
Rc/iation Work Permit ard Maintenance Request to acoceplish testing.
B.
Watch Engineer's permission to start test as indicata$ on this Procedure and/or Maintenance Re< pest.
C.
Implementation of procs5 ural step VII.A.4 will make 1 PER pupp unavailable. Se requirements of hchnical Specification 3.5.A.5 nust be addressed prior to implementing procedure step VII.A.4.
V.
PREX:AUTICNS:
A.
When draining and venting systems that are potentially contaminated, Standard PNPS Ra51ological Protection practices must be followed.
B.
If an abnormal conSition is encountered during the performance of this test, place the system in a safe con $ition, terminate the test ard investigate the cause.
C.
If a contro11e$ leak-off is successfully implemented by this procedure, a flow of approminately 1 GDM will be continuously i
'IP86-859. Bev.1 A4-2
_ =
e F
div:rted to the torus. OperGtions chould be aware of this I
inventory addition and the need for increased water level F
. control for the torus.
i
~ VI. APPAPA1US:
1 A.
Venting bottle B.
Rubber hose, various lengths VII. PRCCEDURE:
STFS A'and B nay be performed in any sequence. Step B may be performed separately for information purposes.
A.
Determine the In-Leakage Rate and Establish Controlled Ieakoff NCTIE:
RHR IiXP 'B' MUST NCfr BE IN SERVICE WHEN cot 00CTING THIS SDT.
THIS ST@ USES THE B IiXP. THE CBOSS-TIE MAY BE CLOSED IF ESSENTIAL TERATION & THE A PHR LOT EKISIS.
1.
Record system pressures as indicated on attachnent A before clearing alarm (if possible) by reducing system pressure, per alarm procedure 2.3.2.1 annurcator B-7.
2.
Close the 10 1001-100D valve if open.
(Note nunber of turns to seat valve)
- 3. Iet system pressure reset to appreminately 150 psi.
4.
Close the 'D' RER suction valve MO 1001-7D an inter 1cck will prevent inadvertent pump start when there is no suction path.
5.
Attach hosing to the 3/4 inch vent connection that is on the pump side of valve ID 1001-1000 and connect other and to vent bottle.
6.
Route petcock drain hose to floor drain, open the petcock.
7.
Open the vent valves 10 1001-340D and 341D.
8.
Establish a leak off flow rate to maintain a system pressure of 150 psig by slightly opening valve 101001-10CD note the nunber of turns open.
9.
When preesure has stabilire5 at approx.150 psig close the petcock at the bottom of the vent bottle. Time the interval required to fill the bottle from the 1 gallon mark to the 5 gallon mark to neasure the in-leakage. Record the time to fill the 4 gallons. A time of less than 4 minutes is not acceptable per the acceptance criteria.
7P86-85-3 Rev. 1 A4-3 i
L L
- +
I 10.
If the acceptance cri'teria is set, tie wrap the 1001-1000 in its present position. If the acceptance criteria is not net inform the Watch Engineer.
- 11. Close the ID 1001-340D ard 341D valves and replace the vent cap. Double verification is required.
- 12. Be-open MO 1000-7D with the control switch.
B.
Corduct a Temperature Survey of MUt Piping 1.
Temperature sensitive monitors (self adhesive strips) bue been installed on the Mut system piping at the locations as shown on Attehnent B.
Se indicators are locates below the thermal insulation on the piping. On rigid insulation, a block of the insulation sust be removed prior to reading the temperature indicators. Ch blanket type insulation, the insulation must be pushed aside in order to read the tenpera-ture irdicators. Insulation must be restored to the as-found cordition af ter each temperature reading is taken.
Check the temperature in51cators for change ard record the
.tsperatures on Attachment B.
2.
Use a portable temperature monitoring device to survey RHR systan piping ard record the location and present pipe wall temperature on Attachment B.
3.
Record Rut system pressures in locations irdicated by Attach-sent B.
4.
Forward the temperature survey data cheet to the STA on shift for further disposition by NED.
VIII. fCCEPINCE CRI'ERIb A.
his procedure will be suitable for controlling systesn pressure with system inleakage rates of 1 gpm or less. If system inleakage is greater than 1 spm, a controlled shutdcwn should be initiated and a NED Disposition should be obtained.
B.
Ccapare pipe wall temperatures outboard of tre normally closed isolation valve (i.e. 1001-28A or 1001-22) to the saturation temperatures at the pressure indicated on the local guages (PI 1001-80A or 1001-80B). If pipe wall temperature is equal to or within 15 F of saturation temperature steam void formation may be present. Inform the watch ergineer of this cordition, and request a3ditional aid from NED.
'IP 86-85-4 Rev 1 A4-4 L
)
- 7., _7,.
.g.
s; h
IX. -
N A. RRR Makage Assessment Setup and Data' Sheet
~
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, 3:.
4
' B. MHR Temperature Sarvey Data Sheet n
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i
.a W 86-85-5 Rev 1 A4-5
W 86-85 RER Leakag2 Assessment Set-Op c.M D2ta Sheet
~
Watch Engineer's permission granted to perform this proedure:
NWE date/ time
/
B3)IINENT STICHES FOR xtruuCE:
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- 1. Initial Corx31tions:
i
'this investigation includes data for RER loops A and B.
'Ib read loop pressures, slowly open the instrunent root valves to obtain the readings, then close the root valves when the readings have been taken.
i INS'mDMENT FIED IiCATION READI!G (psig)
CICSED BY PI 1001-30A in CIO shop west wall PI 1001-30B outside B RHR valve room PI 1001-8GA A BRR Quadrant at rack 2259 PI 1001-80B B RER Quadrant at rack 2262 Instrument root valves verified closed by:
date
- 2. Ieakage Assessment / Controlled Leakoff:
- Close valve 101001-100D (if @en) record the nunber of turns to seat the valve:
turns Let RER system pressure rise to approx.150 psig.
T 86-85-Al Rev. 1 A4-6 t
u_____________________
t.
l
'IP 86-85 RER Isakaga Asressment Set-Up and Data Sheet (Cont.).
]
Mm punp 'D" suction valve MO 1001-7D closed. Verified by:
L Attach hose from venting bottle to the 3/4 irch vent downstseam of B0 1001-100D. Performed by:
Venting bottle petecck drain bose routed to floor drain, petcock opened.
Performed by:
l
- 3/4 inch vent valves HD 1001-34(D, 341D crrked open.
Perfomed by: _
Determine leakage rate by throttling HD 1001-100D and maintaining approx.
150 psig on guage PI 1001-80B or test guage installed at HD 1001-341D.
When pressure has stabilized, close ventirn bottle petcock.
Measure the time require 3 to fill the venting bottle from the lower mark to the upper raark. ' Ibis differerce is 4 gallons. Record the fill tine anS the set pressure.
Fill time minutes: seconds Set Pressure psig Coupute leakoff rate : 4 gallons / time (minutes)=
opm Acceptance criteria is 1.0 gpm or less ( or a fill time of 4 min or nere)
Criteria met
_ or Not met determined by If the acceptance criteria is set tie wrg the ID 1001-100D valve securely in its present position.
BE AWARE 'IEAT KEEP FIII SYSTEN WA'IER IS BEItG DIVERIYD PERMANENTLY 'IO 'IEE
'IURJS - AN I! CREASED AWARNESS T TORJS LEVEL IS NECESSARf.
NOIE:
If the acceptance criteria is not net, then inform the Watch Engineer immediately and return the B01001-10CD valve to the closed position. A CatfrRrY.im SatnDOWN SHALL BE INITIATfD MO 'n{E NLCLEAR DGINEERI!G DEPARIMDfr ALER1YD 'IEAT ADDITIONAL EVAIARTION i
IS NEEDED.
- Close the ID 1001-340D an3 341D 3/4 inch vent valves:
closed by:
date verifie$ by:
date Forward this signoff sheet with the Shift Technical A5 visor for additional evaluation. When evaluations are cocplete the Shift AA is to transmit this form to DOC for permanent record.
'IP 86-85-A2 Rev. 1 A4-7
1:
=
74, W 86-85 Attachnnet B Mm System Tenpercture Survey Data Sheet Note:
Temperature sensitive' monitors (self adhesive strips) have been installed on the Mm system piping at the. locations as shown on the sketch below. 'Ihe indicators are located below the thermal insulation.on the piping. On rigid insulation, a block of the insulation nust be removed prior to reading the temperature indi-cators.. Ch blanket type insulation, the insulation nust be pushed l-aside in order to read the temperature indicators. Insulation-nust be restored to the as-found condition after each temperature reading is taken.
Note:
An WP is required for equipment room entry.
LCCATION SKETCH T Mm TD4PEWIURE MONI'ICRS (Note: See data sheet on following page.)
s A RHR IIOP BPJE TIDP A RHR Valve Boom B PJE Valve Room l
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T 86-85-B1 A4-8 I
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1P 86-85 Attachmnet B HER Systen Temperature &lrvey Data Sheet (Cont.)
I Temperature alrvey Instrument Used:
5-Tine of Highest Current Data fe ation Read ina -
hM N had N Taken Bv Al i
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B7 l
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'Ib res5 loop pressures, slowly open the instrument root valves to obtain the readings, then close the root valves when the readings have been taken.
INSIMMEhT FIIID I4CATICN READING (psig)
Tsat ( F)
PI 1001-30A in CIO shop west wall PI 1001-30B outside B RER valve room PI 1001-80A A RHR Quadrant at rack 2259 PI 1001-808 B RER O2adrant at rack 2262 Instrument root valves closed by:
date verified by:
date Compare current pipe wall temperature rem $ings A1, A2 with saturation temperature at pressure of PI 1001-80A.
Coupare current pipe wall temperature readings ' B1, B2 with saturation temperature et pressure of PI 1001-80B.
If pipe wall temperature is eg2al to or within 15 F of the saturation temperature, steam void formation may exist. Inform the Watch Ihgineer that NED assistance is needed for additional evaluation.
1
'IP 86-85-B2 Rev.1 A4-9 i
b________________.._
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,o 1
ATTACHMENT $
Temporary PNPS Procedure TP 86-84 RHR Discharge Piping Venting Procedure
BOSTON EDISON NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure No. TP86-84 s
RHR DISCHARGE PIPING VENTING PROCEDURE List of Effective Pages TP86-84-1 TP86-84-2 TP86-84-3 TP86-84-4 TP86-84-5 TP86-84-6 Attachments None Approw b
N lear Operations Manager
& ~f 0 "OS Date l
Expiration Date b - /8
[
I EM-
)
TP86-84-1 Rev. O
gy The purpsse of'this procedure is to provide instruction to~ station personnel to. vent air from RHR discharge piping utilizing high point vents.
II.
DISCUSSION This procedure was formulated as a result of the RHR Task Force-i investigation into the RfiR pressurization events of April 1986.
This l'
. procedure is part of a commitnent to the NRC in response to confirmatory action letter 86-10, and is intended to meet RHR Task force recommendation C.3.a.
The schedule for performing this procedure is.to perform it once per week from the time of RHR work completion (ie: upon refill of the RHR system after' maintenance for CAL 86-10) for 4 weeks.
Thereafter, this procedure will be performed in conjunction with RHR system surveillance tests 8.5.2.2 and 8.5.3.3.
If this procedure is proven to be effective and "do-able", it will be i
incorporated into existing RHR system procedures.
It should be noted that proper system venting mitigates excessive pressures-due to water hammer events.
III.
REFERENCE MATERIAL A.
' Procedure 8.5.2.1 8.
Procedure 8.5.2.2 C.
. Tech. Spec. 4.5.A.3.d D.
P&ID M-241 E.
PDC 86-30' IV.
PREREQUISITES A.
Radiation Work Permit (specific or extended per HP group) to accomplish venting.
8.
Watch Engineer's permission to start test as tadicated on this procedure.
C.
Suppression chamber water level within Technical Specification Limitations.
D.
Assure RHR Discharge Header how pressure alarms are not in and keep fill valves to RHR are open.
E.
No RHR pump running, or loop to be vented with cross connection closed if other loop is in service.
F.
When system is restored to normal two independent valve checks must be made.
TP86-84-2 Rev. O A5-2
[
.G.
QMIO2958.
~ ~
e Fcur (4) drnin hoses cith applicable connections.
VI.
PRECAUTIONS' A.
When venting systems that are potentially contaminated, standard PNPS Radiological Protection practices must be followed.
l B.
Prior to venting assure no high pressure RHR ALARM's are in.
VII.
PR0,CEDURE' NOTE:
Sections of this procedure may be performed independently and in conjunction with normal surveillance procedures.
A.
Obtain Watch Engineer's permission to start venting procedure as evidenced by his signature on this procedure.
Watch Engineer's permission to start venting:
/
Watch Engineer Date B.
Venting RHR Discharge piping to Torus, 1.
Open 1001-36A/B 2.
Open 1001-34A/B 3.
Allow Flow to Torus for 5 minutes.
4.
Close 1001-34A/B.
5.
Close 1001-36A/B.
C.
Venting "A" LPCI Piping.
1.
Close 1001-29A, check that vent valves HO-1001-329A and HO-1001-328B are closed.
2.
.Take cap off of vent line located in "A" RHR valve room in between 28A & 29A.
3.
Attach hose to vent line and run to dre'.
4.
Open HO-1001-328B inboard valve.
5.
Crack open HO-1001-329A.
6.
Open inlet valves to pressure gauge PI 1001-30A nounted i
l between MO-1001-28A and 29A.
7.
Allow pressure to decay.
(Observe pressure gauge mounted on drain line between MO 1001-29A and H0 1001-26A.)
TP86-84-3 Rev. O AS-3 m_
_ - - - - _ - - - _ - ~ - - - - -,, - - - - -.
.:d 1.
' CAUTION If pressure increases to greater than keep fill. pressure L
(>130 psig) terminate procedure Notify Watch Engineer l:'
that leakage past the MO-1001-29A is suspected.
L p
8.
Open M0-1001-28A.
9.
Reopen-HD-1001-329A.
- 10. -Allow water to drain until no air present.
- 11. Closed HO-1001-329A and 1001-328B.
L 12.
Disconnect hose and cap line.
- 14. Close MO-1001-28A.
- 15. Open HO-1001-29A.
D.
Venting "B" LPCI Piping 1.
Close MO-1001-29B, check vent valves HO-1001-372 and HO-1001-420 are closed.
2.
Take cap off of vent line located in "B" RHR valve room in between 288 & 298.
3.
Attach hose to vent line and run to drain.
4.
Open HO-1001-372 inboard valve.
5.
Crack open HO-1001-420.
6.
Open inlet valves to pressure gauge PI 1001-30B mounted between MD-1001-28B and 29B.
7.
Allow pressure to decay.
(Observe pressure gauge mounted on drain line between MO 1001-29B and MO 1001-28B on outside of "B" RHR valve room on the north wall.)
8.
Reclose HO-1001-420 and observe pressure.
CAUTION If pressure increases to greater than keep fill pressure
(>130 psig) terminate procedure, notify Watch Engineer that leakage past the MO-1001-29B is suspected.
TP86-84-4 Rev. O AS-4
~
~ ~~ ~ - - - -
t Open MD-1001-288.
8.
i 9.
Reopen HO-1001-420.
l'O.. Allow' water to draib untti no air present.
- 11. Close HO-1001-372 and HO-1001-420.
- 12. Disconnect hose and cap Itne.
14.' Close MD-1001-288.
- 15. Open MD-1001-298.
E..
Upper Drytell Spray (Loop A).
1.
Insure MO-1001-23A and MO-1001-26A are closed.
2.
High point vents located in Cleanup Heat Exchanger Room, Rx 81dg.
51' level, outboard of 1001-23A. Check that vents are closed.
3.
Take off. cap on vent. Attach hose and rup to drain.
4.
Open HO-198 inboard vent and HO-199 outboard vent.
~
+
Open MO 1001-23A.
5.
Allow water to orain until no air is present. Close MO 1001-23A.
- 6. -
Close vents. Disconnect hose and cap.1tne.
F.
- Lower Drywell Spray (Loop B).
1.
l Insure MD-1001-23B-and MO-1001-26B are closed.
High point vents located in Tip Room, Rx. Bldg. 23' outboard of 2.
1001-238. Check that vents are closed.
3.
Take off cap on vent.
Attach hose, and run to drain.
4.
Open 120A inboard and 120B outboard. Open M0'1001-238.
5.
Allow water to drain untti no air present. Close MO 1001-238.
6.
Close vents, disconnect hose and cap itne.
VIII. SYSTEM RESTORATION A.
Verify the system is returned to normal by at least two independent valve lineup checks.
B.
Log conduct of this venting procedure in Station Operating tog.
TP86-84-5 Rev. O AS-5
e-gu.
g ousuemmuu A.
Satisfactory completion of venting' procedure.
'B.
Ifpressurein-stepVIIIC.7orVII.D.7increasestogreaterthan.
keep fill pressure at approximately 130 psig, then leakage past the closed inboard isolation valve MO-1001-29A or B is suspected.
Take the following action:
1.
Check the temperature of RHR piping at installed temperature surveillance points.
2.
Check operating loc book for frequency of the "RHR Discharge or Shutdown Cooling Suction High Pressure" Alarm (annunciator B-7 on panel 903 left).
3.
If annunciator B-7 frequency is less than once per.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, establish normal lineup for RHR-l.PCI and monitor for further valve leakage.
4.
If annunciator B-7 frequency is greater than once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, file an F & M Report and initiate procedure TP 86-85 to determine the in-leakage rate.
X.
ATTACHMENTS None i
TP86-84-6 Rev. O j
i As-6
4,:
ATTACHMENT 6 Sefety Evaluation SE 1959 Establishment of a Controlled Leak in the RHR System per Temporary Modification 86-20 l
l l
Safety Evaluation NED Proposed Change" No.:l969 b o l
$AFETY EVALUAfl0N PILGRIM NUCLEAR POWER STATION Rev. No.
?;; ';#
System Calc.
Initiator:
Dept:
Group:
No.:
Name:
No.:
Date:
3%bJ Me.O
'PEAPE. Tf4*8(r20 ERR.
Wl M9
(,f g3 gg -
40.0 EsLituk,J el a--
Description of Proposed change, test or experiment:
eJil r o t.m em h t..I ter A L h 1 tuft <a tte w b e - h n ac urag 4
L - 24.
SAFETY EVALUATION CONCLUSION $:
The proposed change, test or experiment:
(.4 Does Not ( ) Does increase the probability of occurrence or consequences of an accident or malfunction of equipment important to 1.
safety previously evaluated in the FSAR.
(6-(Does Not ( ) Does increase the possibility for accident or malfunction of a different type than any evaluated previously in the 2.
FSAR.
( f Does Not ( ) Does decrease the margin of' safety as defined in the 3.
b9 sis for any technical specification.
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( ) Not Recommended (4 F.ecomende SE Performed by --
A O1'//
Date M. /f,/At 1
Rev. 2 Exhibit 3.07-A Sheet 1 of 3 A6-1 h*
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pussw.t yf will aff a % opu. In. to be aleefel (.kg x 6m. a +kt sin cersm.) {o av. iata.n vk lea Q pas (. % M.o to0s-it & asd M& val0cS. %
.i 6 Pm lu k is Ic 55 h h gui%.fc4f air 1.ca.k9 oJ(,wed b locpg5o Calculak:n M-249 Re v.o Apped'x 7 kg a. Acb ( approylmafc(g 3.4 v>cti e t A c=Icu.lal<A wafu y, o ns At h41.o Gem it k. <a.f c Att3 leaka y secqh.nce cei k ia kaseL on % dals. okfnined durig eskq f
c of L 6 RHR loop. In fa.ch,6's mods'Oca.Nen prov.' des am imprend meus of assessim cenkkin+k in+p'h. h. add;.bw, % wki-g U isola.km f%*c hws of iooi-zs e a to 6.ac wf bem affeckl h %s cA2wy.
h leak o@ line. will hut. w impact m Rtt4. pq Snally 3 opvah% ss %\\ ud..$ca.h'en. caly proddes, o leak fa.dk PW ncia4y EluA vake. Lu.I ut A q ihetf, All pq d
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Safety Evaluation i
No. :,L Y)* PPS*) Roa\\
SAFET.Y EVALUATION PILGRIM NUCtEAR POWER STATION Rev. No. Z h
A.
APPROVAL (14) ( 4 This proposed change does not involve a change in the Technical Specifications Ref.10CFR50.59(c).
l (14) (v1 Thisproposedchange,testorexperimentdoes()doesnot(g involve an unreviewed safety question as defined in 10CFR, Part 50.59(a)(2).
(15) (( This proposed change involves a change to the FSAR per 10CFR 50.71(e) and is reportable under 10CFR50.59(b).
4 (15) ( )
Coitrnents:
(16)
The safety evaluation basis and conclusion is:
h
( [ Approved
() Wot Approved 74e/e M/1[g (17)
Discipline Group Leader /D' ate Supporting Discipline Group Leaoer/Date 8.
REVIEW APPROVAL (18) ( ) Consnents:
(19)
U N
M/hD O
581A Grouo leader /Daie '
L.
ORC WLVIEW (20) ( ) This proposed change involves an unreviewed safety question and a requests for authorization of this thange must be filed with the Directorate of Licensing, NRC prior to implementation.
(20)'( d This proposed change does not involve an unreviewed safety question.
Date J//C,/N (21)
(21) ORC Chairman S
U U
(22) ORL Meeting Number _J4 - PO cc:
Exhibit 3.07-A Rev. 2 Sheet 2 of 3 A6-3
i l
o PILGRIM STATION FSAR REVIEW SHEET
References:
k O N I
Date:
I9 N tev. No.:
Safety Evaluation:
Support a change List FSAR test, diagrams, and indices affected by this change and corresponding FSAR revision.
Revision to affected FSAR Section is shown on:
Affected FSAR Preliminary Final Section _ _ _
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PRELIMINARY FSAR REVISION (to be completed at time of Safety Eval preparation),
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/Date: 100 h v xeneweo op.
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- /Date:
Approved by: _ 1 FINAL FSAR REVISION (Prepared following operational turnover of r (1) systems structures of components for use at PNPS).
10 ate:
/Date:
__ Reviewed by:
Prepared by:
Attach completed FSAR Change Request Form (Refer to NOP).
(1) l Rev. J.
Exhibit 3.07-A Sheet 3 of 3 A6-4
Safety Eva_1.T4ation No.:
14 5 SAFETY EVALUATION WORK SHF.ET~
Rev. No.
l System Structure Component Failure and Consequence Analyses.
A.
Structure Component Failure Modes Effects of Failure Coments System 1.
bd 2.
3-s General Reference Material Review REGULATORY CALCULATIONS FSAR DESIGN SPECS PROCEDURES GUIDES STAND SECT 10_N PNPS TECHNICAL SPECS.
a.sA T S. 3.5. A M-tm w.O
/Oc a so A n. I 31A t.7.I3 / 8.9. I.C Asmep py rr For the proposed hardware change, identify the failure modes that a For each likely for the components consistent with FSAR assumptions.
B.
failure mode, show the consequences to the system, structuresf Especially show how the failure (s) affects the assigned safety basis (FSAR Text for each system) or plant saf ety f unct components.
Chapter 14 and Appendix G).
Dateblh6 Prepared by ~U 1
V It is a requirement to include this work sheet with the Safety NOTE:
Evaluation.
Rev. 2 Exhibit 3.07-C A6-5 i
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l, ATTACHMENT 7 Plan and Schedule Details Regarding Long Term Actions
Additional Ditnils Item C (Spurious Isolation) Reference (C). page 1 According to the most current Long Term Program, the EPIC computer project implementation is scheduled for completion 3/31/87.
That
. completion date is based ~on a September 1986 refueling outage.
Approximately four months after return to power from the outage are required to complete-system acceptance tests a))owing for contingency.
Refueling Outage 7 is being rescheduled to commence in January,.1987.
However, a firm start date and duration have not yet been established.
EPIC completion will be scheduled (4) months after return to power from RFO #7.
Item C.6 Trend Surveillance History of 400 psig Valve interlock for reliabilit.y.
Reference (B). Attachment (4), page 6 of 6 The results of Surveillance Test 8 M.2-2.1.8 cf Pressure Switches 263-52A and $2B for the RHR injection valve opening pe' missive have i
been compiled for the five year period ending in April, 1986. The-switch has always actuated at a 100% rate. The incidence where recalibration.was needed to restore the setpcint to within lechnical Specification limit is 3 occurrences out of 40 (20 tests per switch) or a 92.7% calibration reliability rate. The present calibration frequency is sufficient to assure proper setpoint; therefore, an increase in test frequency is not warranted.
The recommendation of the RHR lask Force Item C.6 has, therefore, been completed by this compilation and analysis.
Special Training plan for Union and Management Operations Personnel Prior to Station Startup Prior to th. Union Operations Personnel resuming watch standing duties they will receive training as outlined in the following l
- schedule, In addition, all Management Operations Personnel, including t
STA's, will receive the following training prior to station startup.
A7-1
'O 6
SPECIAL REQUALIFICATION TRAINING SESSION 11A SCHEDULE
'.T IME :
8:00AM - 8:30AM Revised Training Schedule H. Salfour T. Sullivan 8:30AM - 9:00AM Managenent Changes / Current P. Mastrangelo Plant Status 9:00AM - 12:00PM Plant Modification Update Complete review of "BEC0 R. Woodard Response to NRC Cal 86-10" G. Sherman (includino all related Temocrary Procedures)
Temporary Modifications D. Hughes 86-14, Change feedwater heater 10SB outlet valve, M03480, from seal-in to jog 86-19, Diesel Generator "A" Relaying Modification 12:00AM-12:30PM Lunch 12:30PM - 4:30PM Significant Industry Ever.ts J. Klein SER 37 Premature Critical-ity Due to Control Rods Being Improper ~1y Withdrawn SER 13 Control Rod Mis-J. Klein opera tion SER 18 Diesel Generator D. Hughes Differential Relays Non Seismically Qualified A7-2
L.gi...- ~
' Miscellaneous Events J. Klein R. Woodard Technical Specifications G. Sherman Amendment #94 Current Memos Procedure Review 1.3.34 Conduct of Operations 2.1.1 Startup from Shutdown 2.1.16 N.P.O. Tour
- 2.2.22 R.C.I.C.
2.2.84 Reactor Recirculation System 2.3.2.1 Panel 903 Left 2.4.21 Double ended break of 3" instrument air / nitrogen line in drywell 2.4.31 Reactor basin / spent fuel pool drain dcwn A7-3
_._____.________________________________j